The present invention relates to a nondestructive testing method and apparatus for determining burnup in nuclear fuel rods and assemblies.
Accurate knowledge of fuel burnup and fuel burnup distribution in nuclear fuel assemblies is necessary to manage refueling and reactor operation for optimum plant performance. Average core burnup can be obtained quite accurately from calculations based on power output. Calculation of local burnup in specific fuel rods or assemblies is more difficult and uncertain because local burnup can be influenced by conditions that change the local neutron flux energy distribution or the local temperature. These changes arise from a variety of conditions such as bowing of fuel rods or fuel assemblies, coolant flow restrictions or crud deposition in localized areas on the fuel cladding. Accurate determination of local burnup can serve to determine the proper fuel assembly relocation during refueling to maximize energy extraction from the fuel, and it can also serve to reveal incipient problems which may not as yet have been detected by other means.
Mass spectrometric isotopic analysis is the traditional and highly accurate method for fuel burnup measurement. Unfortunately this is a destructive method that requires the removal of a fuel rod (or rods) from a fuel assembly and shipment to a hot cell facility for isotopic analysis. The method is expensive and time consuming. The results of the analysis cannot be available soon enough for corrective action during a refueling cycle.
Nondestructive burnup determination can be made by a measurement of specific radiation that continues to be emitted from a fuel assembly after removal from the reactor. This radiation can be directly related to the quantity of a specific fission product that is present and through this relation to the quantity of material fissioned (burnup). It is possible to make such measurements on two forms of radiation-gamma and neutron. Gamma ray measurement requires the use of a multichannel gamma ray electronic analyzer and time and personnel at the examination site to make the measurements. Simultaneous analysis of many fuel rod locations becomes difficult and the time required to perform their measurements may affect the critical path on the refueling outage schedule. Measurement of spontaneous neutron flux has similar problems.
Detectors utilizing solid state track recorders, have been built and used to measure non-destructively the actinide content (plutonium, neptunium, curiuml in aged spent fuel assemblies, approximately three years after removal from the reactor. When fission fragments, protons or alpha particles impinge on certain materials commonly called solid state track recorders (SSTR) such as mica, and other materials, they leave well defined sub-microscopic tracks or craters in the surface of the material from which the nature and abundance of the incident irradiation can be determined.
The foregoing fuel burnup determination techniques suffer from one or more disadvantages in that they are either destructive, too complex or time consuming for use on site during refueling, requiring a laboratory environment, or requiring aging of the spent fuel to reduce radioactivity levels prior to characterization.